1. Technical Field
This invention relates to a system and method for evaluating the safety of fueling plans which are planned during scheduled outage etc. of a nuclear reactor.
More specifically, this invention relates to such a system and method for evaluating nuclear reactor fueling plans which can evaluate the safety, quickly and accurately, for the fueling plans in which fuel shuffling within the reactor is performed for the burned fuel, spent fuel is unloaded to be disposed of, and new fuel is loaded.
2. Description of the Related Art
In a nuclear power plant, generally a scheduled outage is performed about once a year, and at that time fuel is exchanged. There are fuel exchange methods in which: (1) all the fuel is temporarily transferred to fuel storage pool, and then the burned fuel, excluding spent fuel, and new fuel are loaded at the position according to fuel loading patterns determined by the reactor core design; or (2) spent fuel is unloaded from the reactor core, and the burned fuel is moved within the reactor and installed to preferably give a target fuel loading pattern, and then new fuel is loaded.
According to the former method (1), as the fuel is transferred to a fuel storage pool, the fuel is taken out of the reactor core, and the neutron source is in the process of decreasing accordingly, and if the design conditions ensure subcriticality of the reactor core before and after the fuel exchange, subcriticality during the fuel transfer will be ensured.
However, in this method, because the fuel is moved assembly by assembly with a refueling machine, the time required for the fueling is longer compared to the latter fuel exchange method.
According to the latter method (2) on the other hand, because the fuel to be recycled is moved within the reactor, the time required for the fueling is shorter compared to the former method (1).
In this method, however, high reactivity fuel could become concentrated locally within the reactor in a certain fueling step during the fueling, and it is thus necessary to plan fueling procedures so that the reactor core subcriticality can be ensured even if such high reactivity fuel becomes concentrated locally.
Reduction of a scheduled outage period at a nuclear power plant leads to improvement in the operating rate of the plant and is economically very important.
The latter fuel exchange method (2) is preferable in terms of reduced time taken for fueling particularly for a reactor core with a large number of fuel assemblies to be loaded, but presents problems in terms of time taken for the evaluation of reactor core subcriticality at each fueling step.
Note that generally, subcriticality of a boiling water nuclear reactor is evaluated through reactor shutdown margin calculations by a reactor core simulator. The term “reactor shutdown margin” indicates the degree of subcriticality that can be ensured even if a single control rod with maximum control rod worth, of the rods that are being inserted, is withdrawn.
A “control rod worth” of a control rod is the difference between the effective multiplication factor of a reactor core in which the control rod is being inserted and that of the reactor core from which the control rod is being withdrawn. However, a case of withdrawal of a single pair of control rods which can be withdrawn simultaneously may be considered in some cases, depending on the structure of the control rod drive mechanism. In the present application, such pair of control rods is defined as a “control rod pair”.
In the present application, the term “control rod worth” not only means the control rod worth of a single control rod in a case that the single control rod is withdrawn but also includes the control rod worth of a single pair of control rods in a case that the single pair of control rods is withdrawn.
With respect to physics models of a reactor core simulator used in the current reactor core designs, there are a 3-dimensional 3-group model, a 3-dimensional modified 1-group model, a 2-dimensional 3-group model, a 2-dimensional modified 1-group model, etc.
A 3-dimensional model is a physics model that simulates a reactor core by partitioning the reactor core 3-dimensionally (xyz directions) with mesh and performing nuclear and thermal hydraulic coupling calculations.
A 3-group model is a physics model in which neutron flux is divided according to the energy conditions into thermal neutron flux, resonance region flux, and fast neutron flux, each of which provides an independent variable.
A modified 1-group model is a physics model that simulates a reactor core state by obtaining the ratio of thermal neutron flux and fast neutron flux with the fast neutron flux as an independent variable.
A 2-dimensional model is a physics model that simulates a reactor core by partitioning the reactor core, as viewed from above, 2-dimensionally (xy directions) with mesh and performing nuclear and thermal hydraulic coupling calculations. In the 2-dimensional model, however, because a reactor core which is actually 3-dimensional is summarized into 2 dimensions in a nuclear manner, accuracy can be very low in some cases.
That is, in the above physics models of a reactor core simulator, the models that can simulate a reactor core in detail are a 3-dimensional 3-group model, a 3-dimensional modified 1-group model, a 2-dimensional 3-group model, and a 2-dimensional modified one-group model in that order.
In recent years, physics models of a reactor core simulator are in the trend of further subdivision in dealing with new design fuel or mixed oxide (MOX) fuel, and much further subdivided physics models are desired to be applied also in subcriticality evaluation during fueling.
However, subdivided models require longer time for calculations as a result. A 3-dimensional model is a much further subdivided physics model and takes longer calculation time than a 2-dimensional model. Likewise a 3-group model is a much further subdivided physics model and takes longer calculation time than a modified 1-dimensional model.
When reactor shutdown margin calculations are performed for all the fueling procedures (fueling steps) in fuel exchange, the reactor core state in the target fueling step and the reactor core state in the target fueling step wherein a single or a single pair of each control rod or control rod pair which is being inserted is withdrawn from the reactor core, are calculated for all of the fueling steps. Usually the number of fueling steps for fuel exchange is as much as 1.5 times the number of fuel assemblies, and the number of control rods is as much as 0.25 times the number of fuel assemblies. Assuming that the number of fuel assemblies is 872, the number of fueling steps will be about 1,300, and the number of control rods to be inserted will be about 200, and thus the number of cases of calculations will be about 260 thousand. To calculate all the cases with 3-dimensional 3-group model, the time required for the calculations will extend over dozens of days.
However, the time period from nuclear reactor shutdown to fueling is normally as much as 3 days, allowing up to as much as 2 days for fueling procedure establishment, and reactor shutdown margin evaluation requiring dozens of days for calculation time is not acceptable.
For this reason, simple models and 2-dimensional models have been employed as subcriticality evaluation models for a fuel planning system in the conventional art.
Hereinafter explained is a rough flow of a nuclear and thermal hydraulic coupling convergence calculation method in a subcriticality evaluation by a reactor core simulator, using FIG. 7.
First, a nuclear constant is calculated from conditions such as the reactor water temperature (Step 700).
Then, a count C of iterative calculations of neutron flux is initialized to 1 (Step 701).
Then, a neutron source is calculated from given neutron flux and the nuclear constant calculated in Step 700 (Step 702). As for the given neutron flux at this point, the neutron flux that is given as an initial value is used for Step 702 for the first time after the calculations are started, otherwise the neutron flux calculated by the neutron flux calculations of the last processing (C−1) is used (Step 702).
Then, neutron flux is calculated from the neutron source calculated in Step 702 and the nuclear constant calculated in Step 700 (Step 703).
Then, whether the count C of iterative calculations of neutron flux reaches a set iteration count is judged (Step 704), and if the set iteration count is reached, output power is calculated from the neutron flux calculated in Step 703 and the nuclear constant calculated in Step 700 (Step 705).
If the count C of iterative calculations of neutron flux does not reach a set iteration count, then the count C of iterative calculations of neutron flux is added by 1 count, and the processing from Step 702 is repeated.
Then, the output power calculated this time and that calculated last time are compared, and if the convergence judgment is fulfilled, the calculations are terminated, and if not, the processing from Step 701 is performed.